ML20091A248

Forwards Corrected Pages to NEDO-20846,Rev 1, ATWS Study for Monticello Nuclear Generating Plant & Addl Info Re Util

ML20091A248
Person / Time
Site:	Monticello
Issue date:	11/10/1975
From:	Mayer L NORTHERN STATES POWER CO.
To:	Ziemann D Office of Nuclear Reactor Regulation
Shared Package
ML20091A252	List: ML20091A248 ML20091A254
References
NUDOCS 9105140464
Download: ML20091A248 (10)

Text

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NORTHEMN STATE 8 POWER COMPANY MINN S A POLIS. M1 TA 98400 N

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November 10, 1975 ' "

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Mr. D. L. Ziemann, Chief -

Operating Reactors Branch # 2 NOV121975 0

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Division of Reactor Licensing D O k U. S. Nuclear Regulatory Comissit M $. Y Washington, DC 20555 '

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Dear Mr. Ziemann:

MONTICELLO NUCLEAR GINERATING PIANT Docket No. 50-263 License No. DpR 22 Response to August 21, 1975 ATWS totter This letter is in response to your August 21, 1975 letter regarding an Antici-pated Transient Without Scram (AWS) event at the Monticello facility.

The USAEC Technical Report " Anticipated Transients Without Scram for Water Cooled Power Reactors" WASit 1270, September, 1973 identified Monticello as a Class C plant stating that the need for backfitting for this class of plant should be considered on an individual case basis. Your August 21, 1975 letter stated that design modifications should be implemented at Mor.tict '.10 to re-duce the probability or consequences of an ATWS event. As a Class C plant, the analyses required for Monticello by WASH-1270 did not treat AWS as a new design requirement and therefore 'id not involve investigation of acceptable a lte rne t ive s .

As a result of your August 21, 1975 letter, we have had discussions with all licensees of C plants which are similar to Monticello. A joint utility pro-gram is being formulated to evaluate ATWS alternatives for those plants. We expect to present to you in the near future a program, along with a schedule, designed to be compatible with the conditions of the February 28, 197f Staff testimony on ATWS which states "...the probability of occurrence of an AWS event with serious consequences is low enough to satisfy our safety objective today and for tle next few years." (Docket No. 50-263, Supplemental Testimony of Nuclear Regulatory Commission Staff on Contention 11-33, page 93.) We are prepared to work with you to resolve the appropriate backfit considerations on a schedule compatible with tie safety objective stated in WASH-1270. '

Your letter also requested additional information regarding the response of the Monticello plant to an AWS event. Enclosure 1 to this letter provides I

9105140464 751110 IMN l PDR ADOCK 05000263 p PDR

NORT r4ERN CTATED POWER Coh.r*ANY D. L. Ziemann 2 November 10, 1975 i

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some of the additional information requested. Answers to other portions of j

your request should logically follow further AWS analysis. The results of that study may show that certain of the req,.ested infonnation may be ir-or snay need to be modified. Therefore, we will delay our response,

releva where appropriate, until further evaluation of A1VS for C plants has developed the appropriate information.

l A number of minor errors existed in our April 1,1975 AWS submittal which t were identified to our NRC Project Manager in April and May. In order to establish a complete and correct record, we are including Enclosure 2 which provides corrected pages and instructions for inserting the new pages.

Yours very truly, k Ie o, I

L. O, Mayer, PE Manager, Nuclear Support Services UN/Mlly/ deb f

! cc: J. C. Keppler

G. Charnoff J

MPCA J. W. Fe nnan i

Enclosu res I

1 1

Rehulatory Docket File Enclosure 1 ggg The following information repeats the requests for additional information fran the August 21, 1975 letter f ran D. L. Ziemann (trSNRC) to L. O. Mayer (NSP) and provides the respective responses.

Reque s t Nteube r 1 Provide the peak torus water temperature reached during the MSIV closure AIVS. Provide and justify a torus water temp-erature limit. If the calculated temperature exceeds the limit, discuss the plant modifications needed to keep torus water temperature below tie proposed limit. If the peak torus water temperature exceeds 170 r discuss plant modi-fications needed to keep this temperature below 170 F.

Response Number 1 Torus water temperature was calculated and is reported in responses 4,5 and 6 below. As stated in NEDO-20846, the justification of a torus temperature limit is part of a General Electric program currently underway.

Request Number 2 The analysis, as described in the Monticello ATVS report, takes credit for the operator initiating the standby liquid control (SLC) system five minutes af ter the ATWS event. Dis-cuss the indications available to the operator to assure this manual initiation of the SLC.

Response Number 2 There are 5 aspects to be considered in answering this question. They will be considered individually in the order of increasing indication to the reactor o pe ra to r.

A. ATVS event not involving reactor isolation.

B. ATVS event involving reactor isclation.

C. Scram is challenged; t o te '. lack of response.

D. Scram is challenged; partial response but no control rod movement.

E. Operator reaction to scram.

A. ATWS Events Not Involving Reactor Isolation - For purposes of responding to this question, ATVS events have been categorized into two groups. The events having the least bnpact on the plant te those not involving reactor isolation.

If such an event occurred, the oper or would observe the effects of such an event through changes in process variables which he continuously monitors in the control room (reactor power, pressure, system temperatures, radiation levels, etc.)

l He would also observe any change of state such as the automatic initiation of equipment. Significant deviations from steady state conditions are alarmed by the Itghted and audible control room annunciator system which must be acknowledged by the operator to silence the alarm. The plant process computer monitors many of

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) Enclosu re I page 2 1

4 j the same parameters. When an alarm signal is received by the process computer, j an audible alarm alerts the reactor operator and a hard copy of the alcr; c.es-

sage is logged by the alarm typer.

J Ihere is widespread diversity among all of the functions, components and systems which provide the indications of an ATWS event and scram initiation.

i B. ATWS Events Involving Reactor Isolation - The worst of ATWS isolation events I is an MSIV closure as analyzed in report NEDO-20846 wherein it was assumed that l the operator would respond by initiating the SLC system in 5 minutes. In addition to the general indications discussed above, the isolation events are characterized l by the following indications:

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1) Reacter pressure will increase rapidly with indications 1 on a strb chart recorder before the reactor operator.

] Numerous alarms will light and sound almost instantaneously.

2) The neutron flux will spike upscale on the strip chart

! recorders before the reactor operator.

3) The relief valves will open which can be heard by operating j personnel within the plant.

i j 4) The torus pool temperature will be observed to increase.

Each of these parameters will be alarmed visually and audtbly to the control room i operator. This combination of events will immediately te !1 tha operator that a scram should have accompanied this event and he will proceed widt the procedure

) for a scram discussed below, d

C. Scram is Challenged; Total Lack of Rceponse - This situation can occur only in i the unlikely event that a common mode failure affects a specific segment of the i scram system. If the CMF affects sensors of a given function, the scram will be i

initiated by other process variables momentarily. For example, if an MSIV closure occurred along with suf ficient failures of the position switches to pr: vent a scram,

! the reactor pressure sensors and the high neutron flux sensors weald initiate a *

) scram. It is difficult to postulate a CMF which would deprive the operator of the

! specific information that the scram system was challenged. Assuming, for the moment, that there is a total lack of information that the scram system was challenged, the reactor operator would still have the indications discussed in paragraphs A and B which would show the need to initiate the SLCS, D. Scram Challenged; Partial Response but no Control Rod Movement - If the assumed common mode failure affected intermediate canponents between the sensors and the components implementing the scram, the more likely of the very unlikely hypothesized

, ATWS event, one might expect to have additional information available identifying the challenge of the scram system but with failure of rods to move. The scram will j be annunciated directly to the operator and the plant process computer typers will begin printing a sequence of events log and a plant disturbance log. Immediately the operato/s reaction will be to respond according to the scram procedure discussed be-low.

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Enclosure 1 Page 3 E. L, . e sn- R .. tion te Scram - The first .'our steps of the procedure that an operat r (o.iowr a f tcr r. scram are t follows:

. aunce cvs e C s I hsn 4tng system that a scram

.es o a.cred.

• ) Pt s. c che re .ctor mode switch in the Shutdown position

3) .eitfy tatt e t t rods .aSe been inserted by observing the digital i :%irion indication of each control rod displayed or. v; warator console.

i' 'nse rt SFJ' and IRM detectors.

(of these immediate steps i* should be noted that by placing the reactor mode switch in the Shutdown pt ' ion an estomatic interlock acts to again initiate a scram. Also, when the Smi and IRM detectors are inserted and sense a high count rate within the core, they will initiate a scram) After receiving all the in-dications discuosed above, the operator will knos that a scram should have occurred, and in accordance with step 3 of the scram proce6ure he will verit,? . hat the con-trol rods have inserted properly. If he observes that none of the control rods have inserted, he will immediately initiate a manual scram. Because of the de-sign of the scram systen there is a possibility that should suf ficient equipment fail to prevent the automatic scram, the manual scran will still function. In the unlikelp event that the manual scram does not result in control rod movement, the operator would realize that SLC must be initiated to s. tut down the reactor. In the case of a failure to scram following an isolation event, suf ficient infonnation would be avsilable to him within a few seconds upon which he '.uuld base his decision to initiate the SLC system.

The SLC system is initiated by actuating a single keylocked switch in the control room. No further operator action is required. The key to initiate this system is under the control of the shif t supervisor. The shif t supervisor reports to the control room immediately upon the announcement of a scram. It is extremely unlikely that he will be more than a minutes distance from the control room. The largest fraction of a shif t supervisors time is spent in the control room or in an affice adjoining the control room. It is therefore proper to assume that should an isolation event occur with a failure to scram, the operator would be made aware of the situation and have the capability to initiate the S14 system to cor-i rect the situation within five minutes.

1 Request Number 3 In figure 4-3 the relief valve flow oscillates between about 3,000 and 7,000 lb/see from about 30 seconds to 95 seconds i after the ATWS. At about 108 seconds the relief valve flow I

begins to oscillate between 3,000 and 14,000 lb/sec. Explain i this difference in the peak relief velve flow.

Response Number 3 4

please note that tie ordinate of Figure 4-3 has been corrected to read " Flow Rate (1b/sec x 103 )" rather than " Flow Rate (Ib/sec x 104 ),n Figure 4-3 shows the ATWS analysis assuming three plant modifications initiated upon high reactor pressure, recirculation pump trip, feedwater pump trip and ADS inhibit, j _.

Enclosure 1 Page 4 Upon MSIV closure, reactor pressure rises causing the recirculation and feedwater pumps to trip at 4 seconds. The former causes reduction in the flow through the core. Stoppage of feedwater causes the reactor water level to gradually drop (there-by further decreasing the core flow) and also reduces the sub-cooling of the core in-let flow. Both core flow decrease an! core inle; sub-cooling decrease result in in- ,

creased core average voids and therefore decreased core power. There fore, af ter the >

initial pressure and power spikes subside (i.e., after about 30 seconds) the read--

power attains a level of approximately 30% of the initial valve. With MSIV's closed, this power is relieved from the reactor pressure vessel by steam flow through the relief valves which are assumed to operate 'a four groups. At this power level only two relief valve groups are sufficient to relieve all the energy generation. The opening and closing characteristics of the relief valves csuse the relief flow to oscillate between approximately 350 lb/sec and 700 lb/sec (which indicates that the first group of relief valves is open and the second group is cycling).

The reactor level continues to drop due to continued power generation and lack of feedwater flow. When it reaches the low low level, the HPCI system is initiated.

The HPCI flow starts at abo-t 85 seconds and brings water of enthalpy 90 BTU /lbm into the reactor. This relatively cold water increases the core inlet sub-cooling resulting in slightly decreased core voids and increased core power. To relieve the increased power more relief valves are called upon to act. This combined with the dynamic characteristics of the relief valves causes the relief flow to oscillate between 350 lb/sec and 1,400 lb/sec af ter about 100 seconds (indicating that the first group of valves is open and the next three groups aro cycling).

Requests Number 4. 5 and 6 The Technical Specifications present sodium pentaborate solution concentration versus net tank volume in Figure 3.4.1. The con-centration varies from 10.8% to 21.4%. Perform the analysis using each of these concentrations. Justify the use of 13%

as an initial condition listed in Table 3-1 of NED0-20846.

Also justify the poison reactivity worth and specify the re-actor vessel volume.

1 In Section 4.4 of the Technical Specifications a minimum flow I rate of 24 gym for each of the standby liquid control system pumps is listed as a surveillance requireme-P. Perform the analysis using thie value. In Table 3-2 of NED0-20B46 a 28 gpm flow rate per pump is listed. Provide your basis for using this value in your analyis. Specify the total volume of poison injected following the ATWS and indicate the required volume 1

for both hot shutdown and cold shutdown.

j lt is stated that no accounting for possible non-homogeneous mixing was made since this would take a detailed evaluation.

However, GE stated at a meeting with the staff on August 7,1974, that tests were being conducted on borated water mixing phenomena.

Demonstrate that your assumption of uniform mixing is consistent with the experimental data. Otherwien, perform a sensitivity study to show the effects of non-homogeneous mixing of the liquid poison, varying the mixing efficiency from 50% to 100%.

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Enclosure 1 Page 5 j Responses Number 4 5 and 6 i

t A base case calculation for Monticello was provided in NEDO-20846; the inputs and assumptions are listed in that document. Ser.sitivity studies of those parameters j identified in requests number 4, 5 and 6 are summarized in Table 1, below. An estimation of the effect of a lesser reactivity insertion rate on torus temperature I

and pressure is provided in Table 2 based on calculational results of the base case.

. This information, along with miscellaneous requested data presented in Table 3 I

can be used to assess the physical effects of parameters in question.

I Please note that vessel pressure peak, fuel enthalpy peak and cladding oxt.

dation are not affected by change in SLC reactivity ir.sertion rate.

Additional information on this topic may be deemed appropriate for the generic study of ATWS for C plants.

1 Analytical studies treating the primary aspects of the mixing of the sodium j pentaborate solution in the reactor vessel are underway at the present time. These studbs are, at present, expected to be completed by the end of the first quarter of 1976.

j i TABLE 1

Reactivity Insertion Rates Corresponding to Conditions O' ner Than Those Used in the Base Case Condition as Different Corresponding Reactivity Case From Base Case Insertion Rate (-c/Sec)

With 1 SLC Pump With 2 SLC Pumps f

1 None - (Base Case) 1.19 2.38 2 Sodium Pentaborate 0.9886 1.9772 i Concentration = 10. 8%

3 Sodium Pentaborate 1.958? 3.9178 l Concentration = 21.4%

4 SLC Flow Rate = 1.02 2.04 24 gpm/ pump i 5 Mixing Efficiency = 50% 0.595 1.19

, 6 Mixing Efficiency = 75% 0.8925 1.7850 i

T

_ABLE 2 Effect of SLC Reactivity Insertion Rate on the Peak Containment Pressure and Temperature 1

i (SLC Initiation Time = 5 Min. )

Reactivity Containment Containment Insertion Rate Peak Temperature Peak Pressure

-c/Sec. OF (psig) 2.38 1 84 5.1 l 1.19 213.9 6.9 l 0.595 274 48.6

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2 j Enclosure 1 Page 6 i

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! TABLE 3 i

! Miscellaneous Data Requested i -

3 i Reactor vessel volume to normal water level. . . . . . . . e ............ 9,130 ft J

Reactivity required to bring reactor f rom 100%

p owe r io h o t sh u t d own. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1. 7 %

1 Reactivity required to bring reactor from hot

shutdown to cold shutdown........................................... 1.5%

1 j Request Number 7 i

) The staff has submitted to General Electric questions nn NEDO-1 20626 (letter from V. Stello to I. Stuart, Janua ry 28, 1974,

and letter from W. Butler to 1. Stuart, April 9,1975, copies

are enclosed). Respond to the following questions as they apply to Monticello

1, 4, 5, 6, 9, 12, 13, 16, 17, 310.1,,

310.3, and 310.5.

i

Response Number 7 4

I The requests for information referenced above were asked as part of the staff re-

! view of a generic study of Class B BWR plants. Each request will be reviewed in the appropriate perspective as part of the anticipated Claes C generic program and j addressed accordingly, i

l Request Number 8 a

i t Provide the bases for assuming thirty seconds for transport tLme l of the sodium pentaborate solution from the storage tank to the j vessel and for the liquid to become ef fective in the core.

I Response Number 8 j The thirty second transport time is made up of two segments. Approximately half of 1

the time is required to pump the boron solution at rated flow to the sparger im-i mediately below the core. The remaining 15 seconds is an estimate of the tLae

! it takes for the solution to mix with reactor water and become effective in reduc-1 ing the reactivity of the core. Study of boron mixing discussed in the response to requests number 4, 5 and 6 will provide a better basis for transport time, d

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! Enclosure 2 i

i This enclosure transmits corrections to the report submitted l by L. O. Mayer (NSP) to A. Giambusso (USNRC) on April 1,1975 j entitled " Anticipated Transients Without Scram: Study for the

! Monticello Generating Plant, NEDO-20846, March 1975." The l changes include the following:

i l 1. Either destroy the hard cover (which contains i the same infomation as page 1) or change the date j f rom " March 1975" to " Revision 1, May 1975."

J j 2. Destroy the pages of the original report listed 4 below. Replace each page with the respective re-placement page dated "May 1975" which is attached.

j Superceded Pages j i j 4

! 7

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{ 11 1 12 i

14 16

18 1

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NRC D..,TRIBUTION FOR PART 50 DOCKE* .d ATE R I AL

' (TEMPORARY FORM)

CONTROL NO: 12971 FILE:

FROM: NSP DATE OF DOC Minneapolis, Minn. 5540.1 DATE REC'D LTR TWX RPT OTHER L.O. Mayer 11-10-75 11-12-75 XX TO: .

ORIG CC OTHER SENT NRC PDR - XX Mr D.L. Ziemann 40 XX SENT LOCAL PDR CLASS UNCLASS PROP lNFO INPUT NO CYS REC'D DOCKET NO:

XXX -

40 50-263 .

DESCRIPTION:

following: Ltr re our 8-21-15 ler. . .trans LTENCLOSURES: Jar 1 "Addl info re our 8-21-75 l t r r e A'NS . . . . .

Inc1 L " Corrections to the report entitled ATVS Study for the Monticello Plant"(NEDO-

. ~

20846 Class 1 rev. 1, May 1975..."

(40 cys ca enc 1 ree'd) 7,............,

PLANT N AME: Monticello Plant .

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